Neutrons are a fundamental part of any process involving nuclear fission, and thus detection of neutrons is important for radiation protection purposes. Neutron radiation is an ever-present hazard in nuclear reactors. Neutron detectors used for radiation safety must take into account the way damage caused by neutrons varies with energy, and neutron detection techniques may differ depending upon the actual application. Effective neutron detection systems are required to overcome various challenges, such as background noise, high detection rates, neutron neutrality, and low neutron energies. The main components of background noise in neutron detection are high-energy photons (which are not easily shielded), and alpha and beta particles (some of which can be prevented by shielding). Photons are the major source of interference in neutron detection. Unfortunately, both photons and neutrons register similar energies after scattering into a detector from the target, and are thus hard to distinguish from one another. Another challenge is that since the detector typically lies in a region of high beam activity, it is continuously hit by neutrons and background noise at overwhelmingly high rates. This can obfuscate the collected data, since there is extreme overlap in measurement, and separate events are not easily distinguished from each other. It is thus necessary to keep detection rates as low as possible and use a detector that can keep up with high detection rates to yield coherent data.
Neutrons are generated through spontaneous fission, induced fission, or alpha particle induced (α,n) reactions. Because neutrons have mass but no electrical charge, they cannot produce ionization in a detector and, therefore, cannot be detected directly. Detecting neutrons requires an interaction of an incident neutron with a nucleus to produce a secondary charged particle that can itself be detected. The presence of emitted neutrons is thus deduced from the presence of neutrons of such secondary charged particles.
The energy distribution of fission (spontaneous or induced) neutrons is very different to that of (α,n) neutrons, and can thus be used to help determine the source of neutrons. However, it is generally not possible to use simple energy discrimination to distinguish neutrons from different sources because a measurement consists of both cosmic induced neutrons that cover all energies, as well as those from any unknown source of interest. To improve the analysis process, a characteristic time distribution difference between (α,n) neutrons and fission neutrons is analyzed. Fission neutrons typically produce multiple neutrons (e.g., two or three neutrons) simultaneously, whereas (α,n) neutrons are produced individually and randomly. Coincidence counting techniques can thus be used to distinguish fission neutrons from random (α,n) neutrons.
Neutron detection and counting techniques are used to perform non-destructive assays (NDA) of pure samples of plutonium and uranium. Neutron coincidence counting is used to separate the time-correlated fission neutrons from the random, uncorrelated neutrons to determine the fissile mass of the sample. Multiplicity counting is required to analyze impure samples, such as mixed-oxide scrap. Plutonium in bulk form and in waste generates neutrons from spontaneous fission, (α,n) reactions, and induced fission events caused by primary neutrons. Neutron-pair correlation provides the necessary information to determine the spontaneous fission rate and hence the mass of Pu present in a sample, if the isotopic composition is known. The ratio of the (α,n) reaction rate to the spontaneous fission neutron emission rate may be calculated. Coincidence counting requires the effective number of neutron singlets and the effective number of neutron doublets to solve for two unknowns. Multiplicity counting involves the counting of correlated triplets also. With the three quantities (singlets, doublets, and triplets), it is possible to determine three unknowns, such as the spontaneous fission rate, the (α,n) reaction rate, and the detection probability; or the spontaneous fission rate, detection probability, and the neutron multiplication factor. Higher order multiplicity counting is also possible assuming the data is collected in a way to contain the needed information.
A current standard approach to neutron multiplicity counting is through the use of a shift-register sliding word that is gated and counted repeatedly. This usually gives data for one gate width, which is set to correspond to the neutron lifetime. A shift-register is a single-input device where pulses can pile up and be lost. This data loss presents a significant disadvantage for current shift-register based detection systems. Another approach to multiplicity counting is a list mode data acquisition system in which every pulse event is stored in memory. In this system, every pulse is assigned a time-tagged value and stored as a word. The volume of data that accumulates can be on the order of many gigabytes if the objective is a non-destructive assay. The disadvantage of this type of system is that a large quantity of data is required to minimize statistical errors, thus requiring massive amounts of system memory.
It is therefore desirable to provide a neutron multiplicity counter utilizing multiple gates, with different definitions of the gate and counting approaches, and with a parallel architecture that reduces pulse pile up dead time.
In general, multiplicity counters are readily used in conjunction with various types of neutron detectors, and detection hardware refers to the type of neutron detector used and the electronics used in the detector. For example, the most common present detector type is the scintillation detector. The detector hardware defines key experimental parameters, such as source-detector distance, solid angle and detector shielding. Detection software consists of analysis tools that perform tasks such as graphical analysis to measure the number and energies of neutrons striking the detector.
Detectors that rely on neutron absorption are generally more sensitive to low-energy thermal neutrons, and are orders of magnitude less sensitive to high-energy neutrons. Scintillation detectors, on the other hand, have trouble registering the impacts of low-energy neutrons. Although it is sometimes facilitated by higher incoming neutron energies, neutron detection is generally a difficult task, and improving scintillator design has been an ongoing process in the industry. Original scintillation detectors were improved with the advent of the PMT (photomultiplier tube), which gives a reliable and efficient method of detection since it multiplies the detection signal tenfold. Even so, scintillation design has room for improvement as do other methods of neutron detection, other than scintillation. For example, gaseous ionization detectors can be adapted to detect neutrons. While neutrons do not typically cause ionization, the addition of a nuclide with high neutron cross-section allows the detector to respond to neutrons. Nuclides commonly used for this purpose are boron-10, uranium-235 and helium-3. Further refinements are usually necessary to isolate the neutron signal from the effects of other types of radiation. As elemental boron is not gaseous, neutron detectors containing boron use boron trifluoride (BF3) enriched to 96% boron-10 (natural boron is 20% B-10, 80% B-11).
It is further desirable, therefore, to provide a detection system that effectively detects neutrons by adequately compensating for background noise, high detection rates, neutron neutrality, and low neutron energies. Also desired is a system that preprocesses neutron data into small data files in real time, and reduces processing time required for gigabytes of list mode data.
It is yet further desirable to provide a digital data acquisition unit that collects data (e.g., neutron multiplicity data) at high rate and in real-time preprocesses large volumes of data into directly useable forms.